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JAEA Reports

Prototype fast breeder reactor Monju; Its history and achievements

Tsuruga Comprehensive Research and Development Center

JAEA-Technology 2019-007, 159 Pages, 2019/07

JAEA-Technology-2019-007.pdf:19.09MB
JAEA-Technology-2019-007-high-resolution1.pdf:42.36MB
JAEA-Technology-2019-007-high-resolution2.pdf:33.56MB
JAEA-Technology-2019-007-high-resolution3.pdf:38.14MB
JAEA-Technology-2019-007-high-resolution4.pdf:48.82MB
JAEA-Technology-2019-007-high-resolution5.pdf:37.61MB

This report summarizes the history and achievements of the prototype fast breeder reactor Monju. The development of Monju started in 1968 as a prototype reactor following the experimental fast reactor Joyo. The development covers all the activity related to the fast reactor; plant design, mockup tests, construction, operation, and plant management. This report summarizes the history and achievements for 11 technical areas: history and principal achievements, design and construction, operation test, plant safety, core physics, fuel, plant system, sodium technology, materials and mechanical design, plant management, and trouble management.

Journal Articles

Demonstration of under sodium viewer in Monju

Aizawa, Kosuke; Sasaki, Koei; Chikazawa, Yoshitaka; Fukuie, Masaru*; Jimbo, Noboru*

Nuclear Technology, 204(1), p.74 - 82, 2018/10

 Percentile:100(Nuclear Science & Technology)

Development of inspection technique in opaque liquid metal coolant is one of the important issues to ensure the safety of Liquid Metal Fast Breeder Reactor (LMFBR). Performance tests of an Under Sodium Viewer (USV), which was developed to detect an obstacle in the reactor vessel (RV) of LMFBR Monju, have been carried out. The ultrasonic sensors and reflectors are located across the core inside of the Monju's RV. The USV can detect an obstacle existing in between the core top and the Upper Core Structure (UCS) bottom by identifying differences of echo signals. This report describes the USV performance tests. In the tests, the reference echo signals under various conditions were accumulated and the signal to noise ratio successfully exceeded the target value. Measured signals clearly differed from with and without an obstacle. These experimental results show the performance of the USV for detecting an obstacle in the specified place.

JAEA Reports

Evaluation of decay heat used for effectiveness evaluations of countermeasures against severe accidents in the prototype FBR Monju

Usami, Shin; Kishimoto, Yasufumi*; Taninaka, Hiroshi; Maeda, Shigetaka

JAEA-Technology 2018-003, 97 Pages, 2018/07

JAEA-Technology-2018-003.pdf:12.54MB

The decay heat used for effectiveness evaluation of the prevention measures against severe accidents in the prototype fast breeder reactor Monju was evaluated by applying the updated nuclear data libraries based on JENDL-4.0, reflecting the realistic core operation pattern, and setting the rational extent of uncertainty. The decay heats of fission products, the actinide nuclides such as Cm-242, and radioactive structural materials were calculated by FPGS code. The decay heat of U-239 and Np-239 was evaluated based on ANSI/ANS-5.1-1994. The calculation uncertainty of each decay heat was evaluated based on summation of uncertainty factors, C/E values of reaction rates obtained in Monju system startup test, and so on. Furthermore, the decay heat evaluation method based on the FPGS90 was verified by the comparison of the results of the decay heat measurement of the two spent MOX fuel subassemblies in the experimental fast reactor Joyo MK-II core.

Journal Articles

Evaluation of feedback reactivity coefficients by inverse kinetics in Monju

Kitano, Akihiro; Nakajima, Ken*

Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.1205 - 1210, 2018/04

The feedback reactivity is taken into account in fast reactor core design especially in order to make the power coefficient negative, which is required to be confirmed in the operation. In the feedback reactivity experiment, the positive reactivity was inserted in the critical state at zero power, and the thermal data, such as reactor power and the R/V inlet temperature, was acquired until the power got stable by the feedback reactivity. In the conventional study, only two critical points in an experiment are available for evaluation of the feedback reactivity coefficients. This method needs three days for evaluation. The advanced method based on the inverse kinetics is newly applied in this work using the more extensive data. It is confirmed that this approach can evaluate the feedback reactivity coefficients in one experiment.

Journal Articles

Irradiation induced reactivity in Monju zero power operation

Takano, Kazuya; Maruyama, Shuhei; Hazama, Taira; Usami, Shin

Proceedings of Reactor Physics Paving the Way Towards More Efficient Systems (PHYSOR 2018) (USB Flash Drive), p.1725 - 1735, 2018/04

Irradiation dependence of the core excess reactivity was investigated for the Monju system startup tests at zero-power carried out in 2010. The excess reactivity basically decreases with the $$beta$$ decay of $$^{241}$$Pu in zero-power operation. However, the excess reactivity little changed in the two month period of the startup tests, which suggests a positive reactivity insertion during the period. The investigated irradiation dependence shows that the positive reactivity increases with reactor operation and mostly saturates by the fission-dose attained during the Monju zero-power operation in a month ($$sim$$10$$^{17}$$ fissions/cm$$^{3}$$). The saturated positive reactivity is equivalent to approximately 47% of the initially accumulated self-irradiation damage recovery assuming the defects were recovered by the fission-fragment irradiation in the reactor operation.

Journal Articles

A Refined analysis on the power reactivity loss measurement in Monju

Taninaka, Hiroshi; Takegoshi, Atsushi; Kishimoto, Yasufumi*; Mori, Tetsuya; Usami, Shin

Progress in Nuclear Energy, 101(Part C), p.329 - 337, 2017/11

The present paper describes the evaluation of the power reactivity loss data obtained in the Japanese prototype fast breeder reactor Monju. The most recent analysis on the power reactivity loss measurement (Takano, et al., 2008) is updated considering the following findings: (a) in-core temperature distribution effect, (b) crystalline binding effect, (c) logarithmic averaging of the fuel temperature, (d) localized fuel thermal elongation effect, (e) updated Japanese Evaluated Nuclear Data Library, JENDL-4.0, and (f) refined corrections on the measured value. The influences of the updates are quantitatively identified and the most precise and probable C/E value is derived together with a thorough uncertainty evaluation. As a result, it is revealed that the analysis overestimates the measurement by 4.6% for the measurement uncertainty of 2.0%. The discrepancy is reduced to as small as 1.1% when the core bowing effect is considered, which implies the importance of the core bowing effect in the calculation of the power reactivity loss.

Journal Articles

Validation and applicability of reactor core modeling in a plant dynamics code during station blackout

Mori, Takero; Ohira, Hiroaki; Sotsu, Masutake; Fukano, Yoshitaka

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 9 Pages, 2017/04

Since safety measures against severe accidents (SAs) such as a long-term station blackout (SBO) are required for Japanese prototype fast breeder reactor Monju, a validation is necessary for the plant dynamics code during SBO. In order to take into account the phenomena in natural circulation: a heat transfer among subassemblies and a flow redistribution, a whole core model has been developed for the plant dynamics code, Super-COPD. This model has been validated by test results of natural circulation in actual facility. In this study, this whole core model was applied to Monju core to evaluate safety measures against SBO, and the pressure loss model of Monju was validated by comparing with results of the plant trip test from the power of 40%. In addition, an analysis was conducted for SBO to investigate the applicability of this model to Monju. The applicability of this model was confirmed by comparing with analytical results using the model without heat transfer between assemblies.

JAEA Reports

Verification of alternative dew point hygrometer for CV-LRT in MONJU; Short- and long-term verification for capacitance-type dew point hygrometer (Translated document)

Ichikawa, Shoichi; Chiba, Yusuke; Ono, Fumiyasu; Hatori, Masakazu; Kobayashi, Takanori; Uekura, Ryoichi; Hashiri, Nobuo*; Inuzuka, Taisuke*; Kitano, Hiroshi*; Abe, Hisashi*

JAEA-Research 2017-001, 40 Pages, 2017/03

JAEA-Research-2017-001.pdf:5.19MB

In order to reduce the influence on a plant schedule of the MONJU by the maintenance of dew point hygrometers, The JAEA examined a capacitance type dew point hygrometer as an alternative dew point hygrometer for a lithium-chloride type dew point hygrometer which had been used at the CV-LRT in the MONJU. As a result of comparing a capacitance type dew point hygrometer with a lithium-chloride type dew point hygrometer at the CV-LRT (Atmosphere: nitrogen, Testing time: 24 hours), there weren't significant difference between a capacitance type dew point hygrometer and a lithium-chloride type dew point hygrometer. As a result of comparing a capacitance dew point hygrometer with a high-mirror-surface type dew point hygrometer for long term verification (Atmosphere: air, Testing time: 24 months), the JAEA confirmed that a capacitance type dew point hygrometer satisfied the instrument specification ($$pm$$2.04$$^{circ}$$C) required by the JEAC4203-2008.

JAEA Reports

Verification of alternative dew point hygrometer for CV-LRT in Monju

Ichikawa, Shoichi; Chiba, Yusuke; Ono, Fumiyasu; Hatori, Masakazu; Kobayashi, Takanori; Uekura, Ryoichi; Hashiri, Nobuo*; Inuzuka, Taisuke*; Kitano, Hiroshi*; Abe, Hisashi*

JAEA-Research 2016-021, 32 Pages, 2017/02

JAEA-Research-2016-021.pdf:5.0MB

In order to reduce the influence on a plant schedule of the MONJU by the maintenance of dew point hygrometers, The JAEA examined a capacitance type dew point hygrometer as an alternative dew point hygrometer for a lithium-chloride type dew point hygrometer which had been used at the CV-LRT in the MONJU. As a result of comparing a capacitance type dew point hygrometer with a lithium-chloride type dew point hygrometer at the CV-LRT (Atmosphere: nitrogen, Testing time: 24 hours), there weren't significant difference between a capacitance type dew point hygrometer and a lithium-chloride type dew point hygrometer. As a result of comparing a capacitance dew point hygrometer with a high-mirror-surface type dew point hygrometer for long term verification (Atmosphere: air, Testing time: 24 months), the JAEA confirmed that a capacitance type dew point hygrometer satisfied the instrument specification ($$pm$$2.04$$^{circ}$$C) required by the JEAC4203-2008.

Journal Articles

Measurement and analysis of feedback reactivity in the Monju restart core

Kitano, Akihiro; Takegoshi, Atsushi*; Hazama, Taira

Journal of Nuclear Science and Technology, 53(7), p.992 - 1008, 2016/07

 Times Cited Count:2 Percentile:57.25(Nuclear Science & Technology)

A feedback reactivity measurement technique was developed based on a reactivity model featuring components that depend on the reactivity coefficients, denoted as reactor power (K$$_{R}$$) and reactor vessel inlet temperature (K$$_{IN}$$). This technique was applied to the feedback reactivity experiment conducted in the Monju system start-up test in May 2010. A thorough evaluation considering all possible biases and uncertainties revealed that the reactivity coefficients can be evaluated with a measurement uncertainty smaller than 3%. The evaluated reactivity coefficients were simulated considering the temperature distribution in the core. The C/E value of K$$_{R}$$ showed good agreement between calculated and measured values within the established uncertainty, and the value of K$$_{IN}$$ was consistent with that reported in a previous isothermal temperature coefficient experiment. The measured and calculated fuel subassembly outlet temperatures also agreed well within 0.2$$^{circ}$$C.

Journal Articles

Validation of decay heat evaluation method based on FPGS cord for fast reactor spent MOX fuels

Usami, Shin; Kishimoto, Yasufumi; Taninaka, Hiroshi; Maeda, Shigetaka

Proceedings of International Conference on the Physics of Reactors; Unifying Theory and Experiments in the 21st Century (PHYSOR 2016) (USB Flash Drive), p.3263 - 3274, 2016/05

The present paper describes the validation of the new decay heat evaluation method using FPGS90 code with both the updated nuclear data library and the rational extent of uncertainty, by comparing the results of the decay heat measurement of the spent fuel subassemblies in Joyo MK-II core and by comparing with the calculation results of ORIGEN2.2 code. The calculated values of decay heat (C) by FPGS90 based on the JENDL-4.0 library were coincident with the measured ones (E) within the calculation uncertainties, and the C/E ranged from 1.01 to 0.93. FPGS90 evaluated the decay heat almost 3% larger than ORIGEN2.2, and it improved the C/E in comparison with the ORIGEN2.2 code. Furthermore, The C/E by FPGS90 based on the JENDL-4.0 library was improved than that based on the JENDL-3.2 library, and the contribution of the revision of reaction cross section library to the improvement was dominant rather than that of the decay data and fission yield data libraries.

Journal Articles

A Scrutinized analysis on the power reactivity loss measurement in Monju

Taninaka, Hiroshi; Kishimoto, Yasufumi; Mori, Tetsuya; Usami, Shin

Proceedings of International Conference on the Physics of Reactors; Unifying Theory and Experiments in the 21st Century (PHYSOR 2016) (USB Flash Drive), p.2610 - 2621, 2016/05

Reactivity loss due to power ascension (power reactivity loss or power coefficient of reactivity) is thus an important design parameter for determining the number of CRs and plutonium content or inventory in the SFR core design, along with the burnup reactivity loss. Measurements on these reactivity losses were therefore performed during the system startup tests in the Japanese prototype SFR Monju in 1995 and analyses have been carried out for several times. The most recent analysis on the power coefficient measurement in Monju was presented by Takano (Takano, et al., 2008). The following latest findings, which have not been taken into account in the past analyses, are available at present and may affect the existing results: (a) in-core temperature distribution effect, (b) crystalline binding effect, (c) logarithmic averaging of the fuel temperature, (d) localized fuel thermal elongation effect, (e) updated Japanese Evaluated Nuclear Data Library, JENDL-4.0, and (f) refined corrections on the measured value. The influences of refining the calculational models and measured value corrections were therefore quantitatively identified in this study by considering all of these new findings. As a result, it was revealed that the analysis overestimates the experiment by 8.1% for the total uncertainty of 5.9%. Therefore, an additional effect, that is the core bowing effect, was considered in the calculation, and the discrepancy was reduced to 2.9%. The possibility of a significant contribution from the core bowing or deformation effect was thus suggested.

JAEA Reports

Investigation and evaluation of the cause about the loss of split pin which was set to pipe support system of the primary cooling system of Monju

Ichikawa, Shoichi; Kawanago, Sho; Nishio, Ryuichi; Wakimoto, Fumitsugu; Fujimura, Tomofumi; Kobayashi, Takanori; Sakamoto, Tsutomu

JAEA-Review 2015-009, 210 Pages, 2015/07

JAEA-Review-2015-009-01.pdf:60.82MB
JAEA-Review-2015-009-02.pdf:63.6MB
JAEA-Review-2015-009-03.pdf:66.2MB
JAEA-Review-2015-009-04.pdf:62.99MB

The loss of the retaining split pins (four pieces) for clevis pin were confirmed at the inspection of the pipe supports in the Monju prototype fast-breeder reactor in May, 2014. The split pins (two pieces) of ROD RESTRAINT and CONSTANT HANGER were fallen off. The split pins (two pieces) of MECHANICAL SNUBBER were broken at both ends of them. As a result of investigation, a dimple pattern was observed in a fracture surface of broken split pin. This observation result showed that fracture morphology is ductile fracture. A reproduction test, whether split pin was broken by loading the external force to the clevis pin, also gave the same fracture morphology. As the result of all cause investigation, the reason of the broken split pins is that the split pins were loaded shearing stress by the external force loaded to the clevis pin axial direction. The result of the cause investigation and a recurrence prevention measure of this trouble was be reported by this report.

Journal Articles

Performance test of under sodium viewer in Monju

Aizawa, Kosuke; Togashi, Yoshinori; Sasaki, Koei; Chikazawa, Yoshitaka; Fukuie, Masaru*; Jimbo, Noboru*

Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.808 - 816, 2015/05

Inspection technique in opaque liquid metal coolant is one of the important issues for the safety warranty of Liquid Metal Fast Breeder Reactor (LMFBR) core. A performance test of Under Sodium Viewer (USV) which was developed to detect obstacles in reactor vessel of LMFBR Monju was carried out. The ultrasonic sensors and reflectors are located across the core inside the Monju reactor vessel. The USV detects the obstacle between the core top and the bottom of Upper Core Structure (UCS) by differences of echo signals. This reports showed the USV performance test in Monju before power operation. In the test, the basic echo signals in various conditions were accumulated and signal to noise ratio met with the design value. Measured signals with and without obstacles showed difference clearly. Those experimental results showed that basic performance of the USV to detect an obstacle between the core and UCS.

Journal Articles

Validation of core cooling capability analysis in Monju during guillotine pipe break at primary heat transport system

Yamada, Fumiaki; Arikawa, Mitsuhiro*; Fukano, Yoshitaka

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05

In sodium-cooled fast reactor, since the coolant does not need to be pressurized, a pipe break due to the internal pressure does not occur physically. For safety margin in Japanese prototype fast breeder reactor (Monju), the guillotine pipe break accident, i.e., loss of integrity (LOPI) has been analyzed as an extreme assumption for beyond design basis accidents (B-DBAs) in the licensing application for the permit. The cooling capability of the core was re-evaluated in this paper during a large-scale, more specifically guillotine pipe break at the primary heat transport system (PHTS) in Monju, newly considering the following latest findings: (a) Experimental data on sodium boiling in fuel assemblies, (b) Actual PHTS pump coast-down characteristics, and (c) Transient burst test data on irradiated fuel claddings. The analysis models were the validated and simulations were re-performed also using the actual Monju data such as the response time to the trip signals, etc. As a result, it was clarified that the ratio of failed fuel claddings does not exceed around 3% of all of fuel assemblies, as in the past licensing analysis. The safety has been reconfirmed to be secured without significant core damage even under an extreme assumption of a double-ended guillotine pipe break at the PHTS in Monju.

Journal Articles

Safety requirements expected for the prototype fast breeder reactor "Monju"

Saito, Shinzo; Okamoto, Koji*; Kataoka, Isao*; Sugiyama, Kenichiro*; Muramatsu, Ken*; Ichimiya, Masakazu*; Kondo, Satoru; Yonomoto, Taisuke

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 10 Pages, 2015/05

JAEA Reports

Development of three-dimensional diffusion and burn-up code HIZER for Monju core management

Kato, Shinya; Shimomoto, Yoshihiko; Kato, Yuko; Kitano, Akihiro

JAEA-Technology 2014-043, 36 Pages, 2015/02

JAEA-Technology-2014-043.pdf:8.94MB

The core management and operation code system aims to perform core management task efficiently by systematic management of data, analyses and edits, which are needed in the reactor core management and operation. The system consists of the five calculation modules: the reactor constant generation module, the neutronic-thermal calculation module, the radiation analysis module, the core structural integrity estimation module, and the core operation analysis module. In these modules, the neutronic-thermal calculation module is based on the dedicated three-dimensional diffusion and burn-up code HIZER. HIZER can execute core calculations easily for specific design specification and operation patterns of Monju, enabling efficient and accurate evaluation of the Monju core characteristics. This report describes its calculation method and validation results.

JAEA Reports

Annual technical report of the prototype fast breeder reactor Monju; 2013

Fast Breeder Reactor Research and Development Center, Tsuruga Head Office

JAEA-Review 2014-030, 138 Pages, 2014/08

JAEA-Review-2014-030.pdf:71.69MB

The prototype fast breeder reactor Monju has accumulated technical achievements in order to establish the fast breeder reactor cycle technology in Japan using the operation and maintenance experience, etc. This annual report summarizes the primary achievements and the data related to the plant management in Monju during fiscal 2013.

Journal Articles

Performance confirmation of Monju failed fuel detection and location system

Morohashi, Yuko; Suzuki, Satoshi

Proceedings of 2014 International Congress on the Advances in Nuclear Power Plants (ICAPP 2014) (CD-ROM), p.624 - 630, 2014/04

Monju has failed fuel detection systems which consist of the delayed neutron (DN) monitoring system, the cover gas (CG) monitoring system, and the failed fuel detection & location (FFDL) system based on the tagging gas analysis. The DN method and the CG method are used to watch the integrity of fuel assemblies continuously. When a fuel failure of pin-hole level is detected by the CG monitoring, the FFDL system starts operation to identify the location of the failed fuel assembly in the reactor core. In case of a larger fuel failure, the DN monitoring system works and the reactor shuts down automatically. The FFDL system collects the tagging gas which migrates into the reactor cover gas from a failed pin. The tagging gas is made of stable isotopes of Kr and Xe. 270 types of isotopic composition are available using 15 types for Kr and 9 types for Xe. Thus, the isotopic composition of the tagging gas can be made specific to each assembly. The assembly containing a failed fuel pin in the reactor core can be identified by analyzing the isotopic composition. The FFDL system is comprised of two tagging gas concentration devices. The device collects and concentrates the tagging gas by adjusting temperature of activated carbon from 110 K to 420 K. The concentration rate is designed to be higher than 200. In the past examination performed, it was confirmed that the concentration rate meets the requirement with a noble gas concentration of 1 ppm. However, the actual noble gas concentration emitted from a failed fuel is assumed to be much lower. In the present study, the performance of FFDL system was demonstrated by measuring low concentration gas of the actual fuel failure level. The sample gas of concentrations ranging 0.1 ppb to 10 ppb was used. As a result, the concentration rate was confirmed to be more than tens of thousands, which sufficiently satisfies the design demand. Moreover, it was confirmed that the concentration dependence of noble gas was small.

Journal Articles

Multidimensional thermal-hydraulic analysis on natural circulation behavior in ex-vessel fuel storage tank of MONJU

Ono, Jun; Mori, Takero; Sotsu, Masutake; Ohira, Hiroaki

Proceedings of ASME 2013 International Mechanical Engineering Congress and Exposition (IMECE 2013) (DVD-ROM), 9 Pages, 2013/11

The severe accident evaluation on the EVST of MONJU has ever been performed by one-dimensional flow network code "Super-COPD". However, it is difficult to predict thermal-hydraulics in the EVST accurately because the fluid in the EVST is driven by natural circulation. Thus we have performed multidimensional thermal-hydraulic analysis in order to clarify the thermal-hydraulic behavior and evaluate the appropriateness of the flow network model. As a result, it was noted that the multidimensionality on temperature and velocity in the EVST was small enough and the flow network model would be almost appropriate. It should be noted that flow resistance of the supporting plates or the heat transfer center of the cooling coils should be set conservatively for the safety analysis.

184 (Records 1-20 displayed on this page)